AMPP Store - Products tagged with 'crack growth rate'

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SCC of Alloy 152/52 Welds Defects, Repairs, Dilution Zones and HAZs in PWR Water

Product Number: ED22-17243-SG

Author: Peter L. Andresen, Chris Wax

Publication Date: 2022

$20.00

Stress corrosion cracking (SCC) has been observed for over six decades in light water reactors structural components, with wide variations in the rate of SCC initiation and crack growth. Newer materials have been adopted in the last three decades, primarily the ~30% Cr Alloy 690 (UNS N06690) and its weld metals, Alloy 52 (UNS W86052) and Alloy 152 (UNS W86152). These materials were initially viewed as immune to SCC, but are now recognized to be susceptibility to SCC, and can exhibit high growth rates in some conditions.

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SCC Of Alloy 82 Weld Metal In BWR Water

Product Number: ED22-17242-SG

Author: Peter L. Andresen, Katsuhiko Kumagai

Publication Date: 2022

$20.00

There is extensive evidence from laboratory data and plant experience of the SCC susceptibility of Alloy 82 weld metal in both BWR and PWR environments. Two international expert panels evaluated laboratory data under PWR conditions and created disposition curves to address the effects of stress intensity factor (K), temperature, and other factors. Another expert panel is creating a disposition
curve under BWR conditions for K, temperature, corrosion potential, impurities, and other factors. Nickel alloy weld metals at lower Cr levels (~15% Cr for Alloy 182 and ~20% Cr for Alloy 82) are more susceptible to SCC than weld metals of higher Cr content (~30% Cr for Alloy 52/152). This paper focuses on on-thefly effects on SCC growth rate of Alloy 82 weld metal in BWR environments of corrosion potential, water
purity and temperature.

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SCC Susceptibility Potential Range for Carbon Steels Used in Nuclear Waste Tanks Determined with a Crack Growth Rate Technique

Product Number: 51324-21154-SG

Author: B.C. Rollins; S. Chawla; J.S. Page; C.L. Girardot; S.T. Campbell

Publication Date: 2024

$40.00

Stress corrosion cracking (SCC) is a documented threat to the integrity of carbon steel radioactive waste storage tanks at the Hanford Nuclear Reservation. A previously published study compared the slow strain rate (SSR) test technique to three alternative SSR methods. That study showed that, regardless of the technique employed, inconclusive results due to subjective aspects of the post-test evaluation are possible. Additionally, the study suggested that the crack growth rate (CGR) technique might reduce ambiguity and more accurately define the critical cracking potential (CCP). The focus of the current study was determining the potential window of SCC susceptibility using the CGR technique as compared to multiple SSR techniques. The results of the study showed the minimum applied potential required to sustain cracking was dependent on whether intergranular SCC was already established. A more noble potential was required to activate intergranular SCC than it was to maintain cracking.

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The Effect of Halides on SCC Susceptibility of Carbon Steels Used in Nuclear Waste Tanks

Product Number: 51323-18830-SG

Author: B.C. Rollins, S. Chawla, J. Gerst, J.A. Beavers, J.S. Page, C.L. Girardot, S.T. Campbell, N. Sridhar

Publication Date: 2023

$20.00

The Hanford Nuclear Reservation contains radioactive and chemically hazardous wastes arising mostly from weapons production, beginning with World War II and continuing through the Cold War. The wastes are stored in 177 carbon steel underground storage tanks, of which 149 are single-shell tanks (SSTs) and the remaining are double-shell tanks (DSTs). The U.S. Department of Energy, Office of River Protection is responsible for retrieving the tank wastes, treating them in order to encapsulate them in glass logs, and then permanently closing the tanks and associated facilities.

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The Effect Of Potassium Hydroxide Primary Water Chemistry On The IASCC Behavior Of 304 Stainless Steel

Product Number: ED22-17415-SG

Author: Michael Ickes, Peter Chou, Kai Chen, Gary Was

Publication Date: 2022

$20.00

The objective of this work was to compare irradiation-assisted stress corrosion cracking (IASCC) growth behavior in simulated pressurized water reactor (PWR) water with pH maintained with LiOH versus KOH. The U.S. nuclear industry is considering changing PWR primary water chemistries to use KOH in place of LiOH, as a means to ensure a stable supply chain and secure cost savings. This experiment will specifically investigate the impact of these alkali ions on the crack growth rate (CGR) and to examine the crack morphologies generated by the CGR experiment.

Products tagged with 'crack growth rate'

SCC of Alloy 152/52 Welds Defects, Repairs, Dilution Zones and HAZs in PWR Water

SCC Of Alloy 82 Weld Metal In BWR Water

SCC Susceptibility Potential Range for Carbon Steels Used in Nuclear Waste Tanks Determined with a Crack Growth Rate Technique

The Effect of Halides on SCC Susceptibility of Carbon Steels Used in Nuclear Waste Tanks

The Effect Of Potassium Hydroxide Primary Water Chemistry On The IASCC Behavior Of 304 Stainless Steel

Association for Materials Protection and Performance